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Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents

Andrews, Nathan A.; Gauntt, Randall O.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of the of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.

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TYPE Other Report YEAR 2017
OSTIDOI

Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

Andrews, Nathan A.; Faucett, Christopher F.; Haskin, Troy C.; Luxat, Dave L.; Geiger, Garrett G.; Codella, Brittany C.

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

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TYPE SAND Report YEAR 2017
OSTIDOI

Differing system response behavior resultant from diverging core degradation models found within the Melcor-Astec crosswalk

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017

Andrews, Nathan A.; Gauntt, R.; Faucett, C.; Belon, S.; Bouillet, C.; Bonneville, H.

This analysis compares the MELCOR results for the first phase of the Modular Accident Analysis Program (MAAP)-MELCOR Crosswalk to Accident Source Term Evaluation Code (ASTEC) results for the same accident scenario. Similar to the original MAAP MELCOR Crosswalk this analysis contains an analyses of system response of both the containment and RPV, core degradation behavior, lower plenum behavior and lower head failure, and finally hydrogen behavior and generation. The accident scenario developed by EPRI and SNL for this analysis is stylized after accident progression of Fukushima Daiichi Unit 1. However, this accident scenario is for the purpose of code comparison and not for Fukushima Daiichi forensic efforts. The behavior of the main steam line isolation valve, control rod drive mechanism, feedwater system, safety relief valve and the isolation condenser behavior were made constant between the two codes. The MELCOR simulation was run to 16 hours, while the ASTEC simulation was run to the point of lower head failure. Key differences in the system response were found to result from differing thermal hydraulic models, how the two codes treat in-vessel core relocation and how the codes treat debris generated. MELCOR treats the core debris primarily as particulate debris, whereas ASTEC treats debris in a single field – “magma” – which often resembles a molten pool. This has significant importance in predicting the total amount of hydrogen generated and the total amount of convective heat transfer away from degraded core materials. Key differences were also found in the total amount of core debris relocating to the lower plenum and then ex-vessel during the scenario with ASTEC predicting significantly more core debris relocating.

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TYPE Conference Poster YEAR 2017
ScopusOSTI

Example of integration of safety, security, and safeguard using dynamic probabilistic risk assessment under a system-theoretic framework

ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal

Kalinina, Elena A.; Cohn, Brian C.; Osborn, Douglas M.; Cardoni, Jeffrey N.; Williams, Adam D.; Parks, M.J.; Jones, Katherine A.; Andrews, Nathan A.; Johnson, Emma S.; Parks, Ethan R.; Mohagheghi, Amir H.

Transportation of spent nuclear fuel (SNF) is expected to increase in the future, as the nuclear fuel infrastructure continues to expand and fuel takeback programs increase in popularity. Analysis of potential risks and threats to SNF shipments is currently performed separately for safety and security. However, as SNF transportation increases, the plausible threats beyond individual categories and the interactions between them become more apparent. A new approach is being developed to integrate safety, security, and safeguards (3S) under a system-theoretic framework and a probabilistic risk framework. At the first stage, a simplified scenario will be implemented using a dynamic probabilistic risk assessment (DPRA) method. This scenario considers a rail derailment followed by an attack. The consequences of derailment are calculated with RADTRAN, a transportation risk analysis code. The attack scenarios are analyzed with STAGE, a combat simulation model. The consequences of the attack are then calculated with RADTRAN. Note that both accident and attack result in SNF cask damage and a potential release of some fraction of the SNF inventory into the environment. The major purpose of this analysis was to develop the input data for DPRA. Generic PWR and BWR transportation casks were considered. These data were then used to demonstrate the consequences of hypothetical accidents in which the radioactive materials were released into the environment. The SNF inventory is one of the most important inputs into the analysis. Several pressurized water reactor (PWR) and boiling water reactor (BWR) fuel burnups and discharge times were considered for this proof-of-concept. The inventory was calculated using ORIGEN (point depletion and decay computer code, Oak Ridge National Laboratory) for 3 characteristic burnup values (40, 50, and 60 GWD/MTU) and 4 fuel ages (5, 10, 25 and 50 years after discharge). The major consequences unique to the transportation of SNF for both accident and attack are the results of the dispersion of radionuclides in the environment. The dynamic atmospheric dispersion model in RADTRAN was used to calculate these consequences. The examples of maximum exposed individual (MEI) dose, early mortality and soil contamination are discussed to demonstrate the importance of different factors. At the next stage, the RADTRAN outputs will be converted into a form compatible with the STAGE analysis. As a result, identification of additional risks related to the interaction between characteristics becomes a more straightforward task. In order to present the results of RADTRAN analysis in a framework compatible with the results of the STAGE analysis, the results will be grouped into three categories: • Immediate negative harms •Future benefits that cannot be realized •Additional increases in future risk By describing results within generically applicable categories, the results of safety analysis are able to be placed in context with the risk arising from security events.

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TYPE Conference Poster YEAR 2017
ScopusOSTI
Results 26–50 of 51
Results 26–50 of 51